Nuclear reactor system having natural circulation of primary coolant

ABSTRACT

A nuclear reactor system that, in one embodiment, utilizes natural circulation to circulate a primary coolant in a single-phase through a reactor core and a heat exchange sub-system. The heat exchange subsystem is located outside of the nuclear reactor pressure vessels and, in some embodiments, is designed so as to not cause any substantial pressure drop in the flow of the primary coolant within the heat exchange sub-system that is used to vaporize a secondary coolant in another embodiment, a nuclear reactor system is disclosed in which the reactor core is located below ground and all penetrations into the reactor pressure vessel are located above ground.

CROSS-REFERENCE TO RELATED PATENT APPLICATIONS

This application is a continuation of U.S. patent application Ser. No.13/577,163, filed Aug. 3, 2012, which is a national stage entry ofInternational Application No. PCT/US2011/023952, filed Feb. 7, 2011,which claims the benefit of U.S. Provisional Patent Application No.61/416,954, filed Nov. 24, 2010, U.S. Provisional Patent Application No.61/333,551, filed May 11, 2010, and U.S. Provisional Patent ApplicationNo. 61/302,069, filed Feb. 5, 2010, the entireties of which are hereinincorporated by reference.

FIELD OF THE INVENTION

The present invention relates generally to nuclear reactor systems, andspecifically to nuclear reactor systems that utilize natural circulationof the primary coolant in a single-phase, such as pressurized waterreactors (“PWRs”).

BACKGROUND OF THE INVENTION

Over recent years, a substantial amount of interest has grown indeveloping commercially viable PWRs that utilize the phenomenon ofnatural circulation (also known as thermosiphon effect) to circulate theprimary coolant to both cool the nuclear reactor and to vaporize asecondary coolant into motive vapor.

CAREM (Argentina) is a 100 MW(e) PWR reactor design with an integratedself-pressurized primary system through which the primary coolantcirculation is achieved by natural circulation. The CAREM designincorporates several passive safety systems. The entire primary systemincluding the core, steam generators, primary coolant and steam dome arecontained inside a single pressure vessel. The strong negativetemperature coefficient of reactivity enhances the self-controllingfeatures. The reactor is practically self-controlled and need forcontrol rod movement is minimized. In order to keep a strong negativetemperature coefficient of reactivity during the whole operationalcycle, it is not necessary to utilize soluble boron for burn-upcompensation. Reactivity compensation for burn-up is obtained withburnable poisons, i.e. gadolinium oxide dispersed in the uraniumdi-oxide fuel. Primary coolant enters the core from the lower plenum.After being heated the primary coolant exits the core and flows upthrough the riser to the upper dome. In the upper part, the primarycoolant leaves the riser through lateral windows to the external region,then flows down through modular steam generators, decreasing itsenthalpy by giving up heat to the secondary coolant in the steamgenerator. Finally, the primary coolant exits the internal steamgenerators and flows down through the down-corner to the lower plenum,closing the circuit. CAREM uses once-through straight tube steamgenerators. Twelve steam generators are arranged in an annular arrayinside the pressure vessel above the core. The primary coolant flowsthrough the inside of the tubes, and the secondary coolant flows acrossthe outside of the tubes. A shell and two tube plates form the barrierbetween primary and secondary coolant flow circuits.

AST-500 (Russia) is a 500 MW(th) reactor design intended to generate lowtemperature heat for district heating and hot water supply to cities.AST-500 is a pressurized water reactor with integral layout of theprimary components and natural circulation of the primary coolant.Features of the AST-500 reactor include natural circulation of theprimary coolant under reduced working parameters and specific featuresof the integral reactor, such as a built-in steam-gas pressurizer,in-reactor heat exchangers for emergency heat removal, and an externalguard vessel.

V-500 SKDI *(Russia) is a 500 MW(e) light water integral reactor designwith natural circulation of the primary coolant in a vessel with adiameter less than 5 m. The reactor core and the steam generators arecontained within the steel pressure vessel (i.e., the reactor pressurevessel). The core has 121 shroudless fuel assemblies having 18 controlrod clusters. Thirty six fuel assemblies have burnable poison rods. Thehot primary coolant moves from the core through the riser and uppershroud windows into the steam generators located in the downcomer. Thecoolant flows due to the difference in coolant densities in thedowncomer and riser. The pressurizer is connected by two pipelines, tothe reactor pressure vessel and the water clean up system.

The NHR-200 (China) is a design for providing heat for district heating,industrial processes and seawater desalination. The reactor power is 200MW(th). The reactor core is located at the bottom of the reactorpressure vessel (RPV). The system pressure is maintained by N2 andsteam. The reactor vessel is cylindrical. The RPV is 4.8 m in diameter,14 m in height, and 197 tons in weight. The guard vessel consists of acylindrical portion with a diameter of 5 m and an upper cone portionwith maximum 7 m in diameter. The guard vessel is 15.1 m in height and233 tons in weight. The core is cooled by natural circulation in therange front full power operation to residual heat removal. There is along riser on the core outlet to enhance the natural circulationcapacity. The height of the riser is about 6 m. Even in case ofinterruption of natural circulation in the primary circuit due to a LOCAthe residual heat of the core can be transmitted by steam condensed atthe uncovered tube surface of the primary heat exchanger.

While the aforementioned PWRs utilize natural circulation of the primarycoolant to both cool the reactor core and heat the secondary coolant,all of these natural circulation PWRs suffer from the drawback that theheat exchange equipment is integrated with and located within thereactor pressure vessel. Such an arrangement not only makes the heatexchange equipment difficult to repair and/or service but also subjectsthe equipment to corrosive conditions. Furthermore, locating the heatexchange equipment within the reactor pressure vessel results inincreased complexity and a potential increase in the number ofpenetrations into the reactor pressure vessel. However, prior to thepresent invention, the location of the heat exchange equipment withinthe reactor pressure vessel was likely deemed necessary to achieve thenatural circulation of the primary coolant in the PWR cycle.

A drawback of other PWRs that exist in the art is the fact that thereactor pressure vessels have penetrations at both the top portion ofthe reactor pressure vessel and at the bottom portion of the reactorpressure vessel. Still another drawback of existing PWRs is the factthat a substantial length of piping and a large number of joints areused carry the primary coolant from the reactor pressure vessel to theheat exchange equipment, thereby increasing the danger of failure due toa pipe break scenario.

BRIEF SUMMARY OF THE INVENTION

These, and other drawbacks, are remedied by the present invention. Anuclear reactor system is presented herein that, in one embodiment,utilizes natural circulation (i.e., thermosiphon) to circulate a primarycoolant in a single-phase through a reactor core and a heat exchangesub-system, wherein the heat exchange sub-system is located outside ofthe nuclear reactor pressure vessel. In some embodiments, the heatexchange sub-system is designed so as to not cause any substantialpressure drop in the flow of the primary coolant within the heatexchange sub-system that is used to vaporize a secondary coolant. Inanother embodiment, a nuclear reactor system is disclosed in which thereactor core is located below ground and all penetrations into thereactor pressure vessel are located above ground. In certain embodiment,the inventive nuclear reactor system is a PWR system.

In one embodiment, the invention can be a natural circulation nuclearreactor system comprising: a reactor pressure vessel having an internalcavity; a reactor core comprising nuclear fuel disposed within theinternal cavity at a bottom portion of the reactor pressure vessel; aheat exchange sub-system located outside of the reactor pressure vessel;a closed-loop primary coolant circuit that flows a primary coolantthrough the reactor pressure vessel to cool the reactor core and throughthe heat exchange sub-system to transfer heat to a secondary coolant;and wherein operation of the reactor core causes natural circulation ofthe primary coolant through the closed-loop primary coolant circuit in asingle phase.

In another embodiment, the invention can be a nuclear reactor systemcomprising: an elongated reactor pressure vessel having an internalcavity containing a primary coolant, the reactor pressure vesselextending along a substantially vertical axis, a major portion of theaxial length of the reactor pressure vessel located below a groundlevel; a reactor core comprising nuclear fuel disposed within theinternal cavity at a bottom portion of the reactor pressure vesselreactor and. below the ground level; the reactor pressure vesselcomprising a primary coolant outlet port located above the ground level;the reactor pressure vessel comprising a primary coolant inlet portlocated above the ground level; a heat exchange sub-system locatedoutside of the reactor pressure vessel and above the ground level, anincoming hot leg of the heat exchange system fluidly coupled to theprimary coolant outlet port and an outgoing cold leg of the heatexchange system fluidly coupled to the primary coolant inlet port; andwherein the major portion of the reactor pressure vessel is free ofpenetrations.

In yet another embodiment, the invention can be a nuclear reactor systemcomprising: an elongated reactor pressure vessel having an internalcavity containing a primary coolant, the reactor pressure vesselextending along a substantially vertical axis; a reactor core comprisingnuclear fuel disposed within the internal cavity at a bottom portion ofthe reactor pressure vessel reactor; a partition dividing the internalcavity of the reactor pressure vessel into a primary coolant riserpassageway and a primary coolant downcomer passageway, the reactor coredisposed within the primary coolant riser passageway; the reactorpressure vessel comprising a primary coolant outlet port in fluidcommunication with a top portion of the primary coolant riserpassageway; the reactor pressure vessel comprising a primary coolantinlet port in fluid communication with a top portion of the primarydowncomer riser passageway; at least one steam generator located outsideof the reactor pressure vessel, an incoming hot leg of the steamgenerator fluidly coupled to the primary coolant outlet port and anoutgoing cold leg of the steam generator fluidly coupled to the primarycoolant inlet port; and wherein the steam generator does not cause anysubstantial pressure drop in a flow of the primary coolant through thesteam generator resulting from an increase in elevation.

Further areas of applicability of the present invention will becomeapparent from the detailed description provided hereinafter. It shouldbe understood that the detailed description and specific examples, whileindicating the preferred embodiment of the invention, are intended forpurposes of illustration only and are not intended to limit the scope ofthe invention.

BRIEF DESCRIPTION OF THE DRAWINGS

The present invention will become more fully understood from thedetailed description and the accompanying drawings, wherein:

FIG. 1 is a schematic of a natural circulation nuclear reactor systemaccording to one embodiment of the present invention.

FIG. 2 is a schematic of an embodiment of a heat exchange sub-systemthat can be used in the natural circulation reactor system of FIG. 1.

FIG. 3A is a schematic top view of a single-pass horizontal steamgenerator in accordance with an embodiment of the present invention.

FIG. 3B is a schematic side view of the single-pass horizontal steamgenerator of FIG. 3A.

FIG. 4 is a side view of a portion of the natural circulation nuclearreactor system of FIG. 1 according to one structural embodiment.

FIG. 5 is an elevated isometric view of a portion of the naturalcirculation nuclear reactor system of FIG. 1 according to one structuralembodiment.

DETAILED DESCRIPTION OP THE DRAWINGS

The following description of the preferred embodiment(s) is merelyexemplary in nature and is in no way intended to limit the invention,its application, or uses.

Prior to discussing FIGS. 1-5 in detail, an overview of one specificembodiment of the inventive natural circulation reactor system, and itsoperation, will be set forth. Those skilled in the art will appreciatethat the overview is directed to one very specific embodiment and thatthe details thereof are not limiting of the present invention in allembodiments. Furthermore, those skilled in the art will appreciate howthe overview applies to the subsequent detailed discussion of FIGS. 1-5.

I. Overview of One Potential Commercial Embodiment

The inventive nuclear reactor system, in one potential commercialembodiment, is a 145 MWe nuclear reactor designed to provide aneconomical and safe source of clean energy from nuclear fission.Strengths of the inventive nuclear reactor system include its inherentsafety and simplicity of operation. The operational simplicity of theinventive nuclear reactor system and the modest outlay required toestablish and commission it will make it possible to deliver the fruitsof pollution-free nuclear energy to the vast mass of humanity around theglobe that does not presently have access to a reliable source of poweror to a robust electrical enemy delivery system. Competitive with largenuclear reactors on a per-megawatt basis, the inventive nuclear reactorsystem is tailored to add generation capacity to the installed baseincrementally with incremental capital outlays. Due to its inherentoperational simplicity, the inventive nuclear reactor system requires aminimal cadre of trained personnel to run the plant. Multiple units ofthe inventive nuclear reactor system can be clustered at one location orgeographically dispersed without a significant increase in theper-megawatt construction cost. Geographical dispersal and undergroundconfiguration serve as natural antidotes to post-9/11 concerns. Themodest power output of the inventive nuclear reactor system makes it aviable candidate source of reliable electrical energy or for providingheating steam to a city or process steam as a cogeneration plant servingan industrial plant.

As a passive small modular reactor of the PWR genre with safety, ease ofmaintenance and superb security, the inventive nuclear reactor system isideally suited to serve as a reliable power source to strategic nationalassets of any country. Design features of the inventive nuclear reactorsystem that speak to its inherent safety and reliability are:

1. Reactor Core Deep Underground

The reactor core resides deep underground in a thick-walled reactorpressure vessel (RPV) made of an ASME Code material that has decades ofproven efficacy in maintaining reactor integrity in large PWR and BWRreactors. All surfaces wetted by the reactor coolant are made ofstainless steel or Inconel, which eliminates a major source of crudaccumulation in the reactor vessel.

2. Natural Circulation of the Reactor Coolant

The inventive nuclear reactor system does not rely on any activecomponents, such as a reactor coolant pump, for circulating the primarycoolant through the closed-loop primary coolant circuit, which includesflow through the reactor core and the heat exchange sub-system. Instead,the flow of the primary coolant through the reactor pressure vessel, thehorizontal steam generators, and other miscellaneous equipment occurs bythe pressure head created by density differences in the flowing water inthe hot and cold segments of the closed-loop primary coolant circuit.The reliability of gravity as a motive force underpins inherent safetyof the inventive nuclear reactor system. The movement of the primarycoolant requires no pumps, valves, or moving machinery of any kind, incertain embodiments.

3. No Reliance on Off-Site Power

Offsite power is not essential for shutting down the inventive nuclearreactor system. The rejection of reactor residual heat during theshutdown also occurs by natural circulation. Thus, the need for anemergency shutdown power supply at the site—a major concern for nuclearplants is eliminated.

4. Assurance of a Large Inventory of Water Around and Over the ReactorCore

The reactor pressure vessel of the inventive nuclear reactor system hasno penetrations in its below-ground portion, which can be the bottom 100feet, which means that the reactor core will remain submerged in a largeinventory of water. All penetrations in the reactor pressure vessel arelocated in the above-ground portion, or top portion, of the reactorpressure vessel and are small in size. The absence of large piping inthe closed-circuit primary coolant circuit precludes the potential of a“large break” LOCA event.

5. All Critical Components Readily Accessible

Both the heat exchange sub-system, which includes the steam generators,and the control rod drive system are located outside the reactorpressure vessel at a level that facilitates easy access, making theirpreventive maintenance and repair a conveniently executed activity. Eachof the steam generators is a horizontal pressure vessel with built-indesign features to conveniently access and plug tubes.

6. Demineralized Water

The primary coolant (which can also be referred to as the reactorcoolant) is demineralized water, which promotes criticality safetybecause of its strong negative reactivity gradient with rise intemperature. Elimination of borated water also simplifies the nuclearsteam supply system (NSSS) by eliminating the systems and equipmentneeded to maintain and control boron levels in the primary coolant. Purewater and corrosion resistant primary coolant loop help minimize crudbuildup in the reactor pressure vessel.

7. Modularity

One can build only one of the inventive nuclear reactor systems at asite, or a large number thereof. Clustering a number of inventivenuclear reactor systems at one site will reduce the overall O&M costs.

8. Long Operating Cycle

The inventive nuclear reactor system will operate for approximately 3.5years before requiring refueling.

9. Short Construction Life Cycle

Virtually all components of the inventive nuclear reactor system areshop fabricated. Site work is limited to reinforced concreteconstruction and a limited amount of welding to assemble the shop-builtequipment and parts. As a result, it is possible to complete theconstruction of one of the inventive nuclear reactor systems in 24months from the first shovel in the ground.

10. Efficient Steam Cycle

A pair of two horizontal steam generators are arranged in series andintegrally welded to the reactor pressure vessel. The efficiency of thepower cycle of the inventive nuclear reactor system, and itscompactness, is further enhanced by superheaters that are integrallywelded to the horizontal steam generators. The superheaters, oneattached to each steam generator, increases cycle efficiency and alsoprotect both the high pressure and low pressure turbines from thedeleterious effect of moist steam.

11. Integral Pressurizer

The design of the reactor pressure vessel incorporates an integralpressurizer that occupies the upper reaches of the reactor pressurevessel. The pressurizer serves to control the pressure in the reactorvessel.

12. Suitable for Water-Challenged Sites

The inventive nuclear reactor system can be installed at sites withlimited water availability, such as creeks and small rivers that areinadequate for large reactors. The inventive nuclear reactor system canbe operated equally well in a water-challenged region by usingair-cooled condenser technology to reject the plant's waste heat. Usingair in lieu of water, of course, results in a moderate increase in theplant's cost.

12. System Parameters in the Safe and Proven Range

The operating pressure and temperature within the reactor pressurevessel is in the proven range for PWRs. Lower core power density thanthat used in large PWRs for improved thermal-hydraulic control (pleasesee table below) and an improved margin to departure-from-nucleateboiling in the reactor core.

Exemplary System Parameters Data Number of fuel assemblies in the core32 Nominal thermal power, MWt 446 Nominal recirculation rate, MLb perhour 5.46 Reactor water outlet temperature, deg. F. 580 Reactor waterinlet temperature, deg. F. 333 Reactor pressure, pounds per sq. inch2,250 Water in the RV cavity, gallons 30,00

13. Minimized Piping Runs and Minimum Use of Active Components toEnhance Reliability and Cost Competitiveness

The amount of piping in the close-loop primary coolant circuit and thesecondary coolant circuit in the inventive nuclear reactor system is theleast of any nuclear plant design on the market, as is the number ofpumps and valves.

14. In-Service Inspection

All weld seams in the primary system including those in the reactorpressure vessel wall are available at all times for inspection. Inparticular, the weld seams in the reactor pressure vessel can beinspected by operating a manipulator equipped in-service inspectiondevice in the reactor well during power generation. Thus, inventivenuclear reactor system exceeds the in-service inspection capabilityexpected of nuclear plants under ASME Code Section XI.

15. Earthquake Hardened Design

Virtually all major equipment in the inventive nuclear reactor systemare either underground or horizontally mounted to withstand strongseismic motions. This includes the reactor pressure vessel, the fuelpool, the reactor water storage tank (all underground) and thehorizontal steam generators, the horizontal superheaters, and thehorizontal kettle reboiler that are floor mounted.

16. Aircraft Impact Proof Containment

The containment structure of the inventive nuclear react system isdesigned to withstand the impact of a crashing lighter plane or acommercial liner without sustaining a thin-wall breach.

II. Detail

Referring now to FIG. 1, a natural circulation nuclear reactor system1000 (hereinafter the “reactor system 1000”) is illustrated according toone embodiment of the present invention. The reactor system 1000generally comprises a reactor pressure vessel 100 and a heat exchangesub-system 200. The reactor pressure vessel 100 contains a primarycoolant 101 that is used to cool the rector core 102 and to heat asecondary coolant within the heat exchange sub-system 200. The reactorpressure vessel 100 is fluidly coupled to an incoming hot leg 201 of theheat exchange sub-system 200 via a primary coolant outlet port 103.Similarly, the reactor pressure vessel 100 is also fluidly coupled to anoutgoing cool leg 202 of the heat exchange sub-system 200 via a primarycoolant inlet port 104. As a result, a closed-loop primary cool tintcircuit 300 is formed through which the primary coolant 101 flows in asingle-phase. As discussed in greater detail below, the flow of theprimary coolant 101 through the closed-loop primary coolant circuit is anatural circulation flow induced by the heat given off by the normaloperation of the reactor core 102.

In certain embodiments, the internal cavity 105 of the reactor pressurevessel 100 is maintained under sufficient pressure to maintain theprimary coolant 101 in a liquid-phase despite the high temperaturewithin the rector pressure vessel 100. In the exemplified embodiment, apressure control sub-system 50 (commonly referred to in the art as apressurizer) is located within a top region of the reactor pressurevessel 100 and is configured to control the pressure of the internalcavity 105 of the reactor pressure vessel 100. The pressure controlsub-system 50 is integral with the removable head 106 of the reactorpressure vessel 100 to prevent line break concerns and to provide a morecompact reactor system 1000. Pressurizers are well known in the art andany standard pressurizer could be used as the pressure controlsub-system 50. In one embodiment, the internal cavity 105 of the reactorpressure vessel 100 is maintained at a pressure in a range of 2000 psiato 2500 psia. In one more specific embodiment, the internal cavity 105of the reactor pressure vessel 100 is maintained at a pressure between2200 psia to 2300 psia. Of course, the exact pressure maintained in theinternal cavity 105 of the reactor pressure vessel 100 is not to belimiting of the invention unless specifically claimed.

The reactor pressure vessel 100 is an elongated tubular pressure vesselformed by a thick wall made of an acceptable ASME material, such asstainless steel. The reactor pressure vessel 100 extends from a bottomend 107 to a top end 108 along a substantially vertical axis A-A,thereby defining an axial length of the reactor pressure vessel 100. Inone embodiment, the reactor pressure vessel 100 has an axial length ofover 100 feet to facilitate an adequate level of turbulence in therecirculating primary coolant 101 from the natural circulation (alsoreferred to as thermosiphon action in the art). In certain otherembodiments, the reactor pressure vessel 100 has an axial length in arange between 100 feet to 150 feet. Of course, the invention is not solimited in certain alternate embodiments.

The reactor pressure vessel 100 generally comprises a domed head 106 anda body 109. The domed head 106 is detachably coupled to a top end of thebody 109 so as to be removable therefrom for refueling and maintenance.The domed head 106 can be coupled to the body 109 through the use of anysuitable fastener, including bolts, clamps, or the like. In theexemplified embodiment, the body 109 comprises an upper flange 110 andthe domed head 106 comprises a lower flange 111 that provided matingstructures through which bolts 114 (FIG. 4) extend to couple the domedhead 106 to the body 109. When the domed had 106 is coupled to the body109, a hermetic seal is formed therebetween via the use of a gasket orother suitably contoured interface.

The body 109 of the reactor pressure vessel 100 comprises an upstandingtubular wall 112 and a domed bottom 113 that hermetically seals thebottom end 107 of the reactor pressure vessel 100. The tubular wall 112has a circular transverse cross-sectional profile in the illustratedembodiment but can take on other shapes as desired. In the exemplifiedembodiment, the domed bottom 113 is integral and unitary with respect tothe tubular wall 112. Of course, in other embodiments, the domed bottom113 may be a separate structure that is secured to the tubular wall 112via a welding or other hermetic connection technique, such as theflanged technique described above for the domed head 106 and the body109. Integral and unitary construct of the domed bottom 113 and the body109 is, however, preferable in certain embodiments as it eliminatesseams and/or interfaces that could present rupture potential.

The reactor pressure vessel 100 forms an internal cavity 105 in which areactor core 102 is housed. The reactor core 102 comprises nuclear fuel,in the form of fuel assemblies, as is known in the art. The details ofthe structure of the reactor core 102 are not limiting of the presentinvention in and the reactor system 1000 can utilize any type of reactorcore or nuclear fuel. The reactor core 102 is positioned in a bottomportion 115 of the reactor pressure vessel 100. In one embodiment, thereactor core 102 has a core thermal power of 400 MWt to 600 MWt duringthe operation thereof.

In one embodiment, the reactor core 102 is comprised of verticallyarrayed fuel assemblies. The spacing between the fuel assemblies isgoverned by the design objective of keeping the reactivity (neutronmultiplication factor) at 1.0 at all locations in the reactor pressurevessel 100. The criticality control in the axial direction is providedby the built-in neutron poison in the fuel rods (called IFBAs byWestinghouse) and possibly by control rods.

A partition 120 is provided within the internal cavity 105 of thereactor pressure vessel 100 that divides the internal cavity into aprimary coolant riser passageway 105A and a primary coolant downcomerpassageway 105B. Both the passageways 105A, 105B are axially extendingvertical passageways that form part of the closed-loop primary coolantcircuit 300.

In the exemplified embodiment, the partition 120 comprises an upstandingtubular wall portion 120A and a transverse wall portion 120B. Thetubular wall portion 120A is an annular tube that is mounted within theinternal cavity 105 of the reactor pressure vessel 100 so as to beconcentrically arranged with respect to the upstanding wall 112 of thereactor pressure vessel 100. As a result, the primary coolant downcomerpassageway 105B is an annular passageway that circumferentiallysurrounds the primary coolant riser passageway 105A. The primary coolantdowncomer passageway 105B is formed between an outer surface 121 of theupstanding tubular wall portion 120A of the partition 120 and the innersurface 116 of the upstanding wall 112 of the reactor pressure vessel100. The primary coolant riser passageway 105B is formed by the innersurface 122 of the upstanding tubular wall portion 120A of the partition120.

The transverse wall portion 120B is an annular ring-like plate that isconnected to a top end of the of the upstanding tubular wall portion120A of the partition 120 at one end and to the upstanding wall 112 ofthe reactor pressure vessel 100 on the other end. The transverse wallportion 120B acts a separator element that prohibits cross-flow of theprimary coolant 101 between the primary coolant riser passageway 105Aand the primary coolant downcomer passageway 105B within the top portion117 of the reactor pressure vessel 100. In essence, the transverse wallportion 120B forms a roof of the primary coolant downcomer passageway105B that prevents the heated primary coolant 101 that exits the reactorpressure vessel 100 via the primary coolant outlet port 103 from mixingwith the cooled primary coolant 101 that enters the reactor pressurevessel 100 via the primary coolant inlet port 104, and vice-versa.Cross-flow of the primary coolant 101 between the primary coolant riserpassageway 105A and the primary coolant downcomer passageway 105B isprohibited by the upstanding tubular wall portion 120A of the partition120.

In addition to physically separating the flow of the heated and cooledprimary coolant 101 within the primary coolant downcomer and riserpassageways 105A, 105B as discussed above, the partition 120 alsothermally insulates the cooled primary coolant 101 within the primarycoolant downcomer passageway 105B from the heated primary coolant 101within the primary coolant riser passageway 105A. Stated simply, onedoes not want heat to transfer freely through the partition 120. Thus,it is preferred that the partition 120 be an insulating partition in thesense that its effective coefficient of thermal conductivity (measuredradially from the primary coolant riser passageway 105A to the primarycoolant downcomer passageway 105B) is less than the coefficient ofthermal conductivity of the primary coolant 101.

Making the effective coefficient of thermal conductivity of thepartition 120 less than the coefficient of thermal conductivity of theprimary coolant 101 ensures that the primary coolant 101 in the primarycoolant downcomer passageway 105B remains cooler than the primarycoolant 101 in the primary riser passageway 105A, thereby maximizing thenatural circulation rate of the primary coolant 101 through theclosed-loop primary coolant circuit 300. In a very simple construction,this can be achieved by creating the partition 120 out of a single solidmaterial that has a low coefficient of thermal conductivity. However, itmust be considered that the material should neither degrade nor deformunder the operating temperatures and pressures of the reactor pressurevessel 100. In such an embodiment, the effective coefficient of thermalconductivity is simply the coefficient of thermal conductivity of thesingle solid material.

In the exemplified embodiment, the low coefficient of thermalconductivity of the partition 120 is achieved by making the partition120 as a multi-layer construction. As exemplified, the partition 120comprises an insulating layer 124 that is sandwiched between two outerlayers 125A, 125B. In one embodiment, the insulating layer 124 is arefractory material while the outer layers 125A, 125B are stainlesssteel or another corrosion resistant material. In certain embodiments,the insulating layer 124 is full encased in the outer layers 125A, 125B.

The internal cavity 115 of the reactor pressure vessel 100 alsocomprises a plenum 118 at the bottom portion 115 of the reactor pressurevessel 100 that allows cross-flow of the primary coolant 101 from theprimary coolant downcomer passageway 105B to the primary coolant riserpassageway 105A. In the exemplified embodiment, the plenum 118 iscreated by the fact that the bottom end 123 of the upstanding tubularwall portion 120A of the partition 120 is spaced from the inner surface119 of the domed bottom 113, thereby creating an open passageway. Inalternate embodiments, the partition 120 may extend all the way to theinner surface 119 of the domed bottom 113. In such embodiments, theplenum 118 will be formed by providing a plurality of apertures/openingsin the partition 120 so as to allow the desired cross-flow.

The internal cavity 105 further comprises a plenum 126 at the topportion 117 of the reactor pressure vessel 100. The plenum 126 allowsthe heated primary coolant 101 that is rising within the primary coolantriser passageway 105A to gather in the top portion 117 of the reactorpressure vessel 100 and then flow transversely outward from the verticalaxis A-A and through the primary coolant outlet port 103.

The reactor core 102 is located within the primary coolant riserpassageway 105A above the bottom plenum 118. During operation of thereactor core 102, thermal energy produced by the reactor core 102 istransferred into the primary coolant 101 in the primary coolant riserpassageway 105A adjacent the reactor core 102, thereby becoming heated.This heated primary reactor coolant 101 rises upward within the primarycoolant riser passageway 105A due to its decreased density. This heatedprimary coolant 101 gather in the top plenum 126 and exits the reactorpressure vessel 100 via the primary coolant outlet port 103 where itenters the heat exchange sub-system 200 as the incoming hot leg 201. Inone embodiment, the heated primary coolant 101 entering the hot leg 201of the heat exchanger has a temperature of at least 570° F., and inanother embodiment a temperature in a range of 570° F. to 620° F.

This heated primary coolant 101 passes through the heat exchangesub-system 200 where its thermal energy is transferred to a secondarycoolant (described below in greater detail with respect to FIG. 2),thereby becoming cooled and exiting the heat exchange sub-system 200 viathe cold leg 202. When exiting the cold leg 202 of the heat exchangesub-system, this cooled primary coolant 101 has a temperature in a rangeof 300° F. to 400° F. in one embodiment. In another embodiment, the heatexchange sub-system 200 is designed so that the temperature differentialbetween the heated primary coolant in the hot leg 201 and the cooledprimary coolant in the cold leg is at least 220° F.

The cooled primary coolant 101 exiting the cold leg of the heat exchangesub-system 200 then enters the reactor pressure vessel 100 via theprimary coolant inlet port 104, thereby flowing into a top portion 127of the primary coolant downcomer passageway 105B. Once inside theprimary coolant downcomer passageway 105B, the cooled primary coolant101 (which has a greater density than the heated primary coolant 101 inthe primary coolant riser passageway 105A) flows downward through theprimary coolant downcomer passageway 105B into the bottom plenum 118where it is drawn back up into the primary coolant riser passageway 105Aand heated again by the reactor core 102, thereby completing a cyclethrough the closed-loop primary circuit 300.

As discussed above, operation of the reactor core 102 causes naturalcirculation of the primary coolant 101 through the closed-circuitprimary coolant circuit 300 by creating a riser water column within theprimary coolant riser passageway 105A and a downcomer water columnwithin the primary coolant downcomer passageway 105B. In one embodiment,the riser water column and the downcomer water column have a verticalheight in a range of 80 ft. to 150 ft., and more preferably from 80 ft.to 120 ft. The vigorousness of the natural circulation (or thermosiphonflow) is determined by the height of the two water columns (fixed by thereactor design), and the difference between the bulk temperature of thetwo water columns (in water the SES and the downcomer space). Forexample, water at 2200 psia and 580° F. has density of 44.6 lb/cubicfeet. This density increases to 60.5 lb/cubic feet if the temperaturereduces to 250° F. The hot and cold water columns 60 feet high willgenerate a pressure head of 6.6 psi which is available to drive naturalcirculation of the primary coolant 101 through the closed-loop primarycoolant circuit 300. A 90 feet high column will generate 50% greaterhead (i.e., 9.9 psi).

As a result of the natural circulation of the primary coolant 101achieved by the water columns and gravity, the reactor system 1000 isfree of active equipment, such as pumps or fans, for forcing circulationof the primary coolant through the closed-loop primary coolant circuit.

In the embodiment illustrated in FIG. 1, the, primary coolant outletport 103 is at a slightly lower elevation (1-3 ft.) than the primarycoolant inlet port 104. However, in other embodiments, the primarycoolant outlet port 103 and the primary coolant inlet port 104 will beat substantially the same elevation (see FIGS. 4 and 5). When theprimary coolant outlet port 103 and the primary coolant inlet port 104are at substantially the same elevation the partition 120 will beappropriately designed. Furthermore, as used herein, the term portincludes mere apertures or openings.

In one embodiment, the primary coolant 101 is a liquid that has anegative reactivity coefficient. Thus, the chain reaction in the reactorcore 102 would stop automatically if the heat rejection path to the heatexchange sub-system 200 is lost in a hypothetical scenario. Thus, thereactor system 1000 is inherently safe. In one specific embodiment, theprimary coolant 101 is demineralized water. All systems and controlsused to maintain boron concentration in the reactor vessel in a typicalPWR are eliminated from the reactor system 1000. Moreover, the use ofdemineralized water as the primary coolant 101 and the existence of thecorrosion resistant surfaces of the reactor pressure vessel 100 helpmaintain crud buildup to a minimum. The reactivity control in thereactor core 102 is maintained by a set of control elements (burnablepoisons) that are suspended vertically and occupy strategic locations inand around the fuel assemblies to homogenize and control the neutronflux.

Referring now to FIGS. 1, 4 and 5 concurrently, it can be seen that amajor portion 130 of the axial length of the reactor pressure vessel 100located below a ground level 400 while a minor portion 131 of the axiallength of the reactor pressure vessel 100 extends above the ground level400. As such, the reactor core 102 is located deep below the groundlevel 400 White the heat exchange sub-system 200 is located above theground level 400. In one embodiment, the heat exchange sub-system 200 isat an elevation that is 80 ft. to 150 ft, and preferably 80 ft. to 120ft., greater than the elevation of the reactor core 102.

The minor portion 131 of the reactor pressure vessel 100 includes a topportion 132 of the body 109 and the domed head 106. The primary coolantoutlet port 103 and the primary coolant inlet port 104 are located onthe minor portion 131 of the reactor pressure vessel 100 that is abovethe ground level 400. More specifically, the primary coolant outlet port103 and the primary coolant inlet port 104 are located on the topportion 132 of the body 109 of the reactor pressure vessel 100 that isabove the ground level 400.

The major portion 130 includes a majority of the body 109 and the domedbottom 113. In certain embodiment, the major portion 130 of the reactorpressure vessel 130 is at least 75% of the axial length of the reactorpressure vessel 100. In other embodiments, the major portion 130 of thereactor pressure vessel 130 is between 60% to 95% of the axial length ofthe reactor pressure vessel 100. In another embodiment, the majorportion 130 of the reactor pressure vessel 130 is between 75% to 95% ofthe axial length of the reactor pressure vessel 100.

The reactor pressure vessel 100 comprises a reactor flange 150. The topportion 132 of the body 109 of the reactor pressure vessel 100 is weldedto the reactor flange 150, which is a massive upper forging. The reactorflange 150 also provides the location for the primary coolant inlet port104 and the primary coolant outlet port 103 (FIGS. 4 and 5), and theconnections to the heat exchange sub-system 200 (and for the engineeredsafety systems to deal with various postulated accident scenarios). Thisreactor flange 150 contains vertical welded lugs to support the weightof the reactor pressure vessel 100 in the reactor well 410 in avertically oriented cantilevered manner (FIG. 1). As a result, thereactor pressure vessel 100 is spaced from the wall surfaces 411 andfloor surface 412 of the reactor well 410, thereby allowing the reactorpressure vessel 100 to radially and axially expand as the reactor core102 heats up during operation and causes thermal expansion of thereactor pressure vessel 100.

Furthermore, the major portion 130 of the reactor pressure vessel 100 isfree of penetrations. In other words, the major portion 130 of thereactor pressure vessel 100 comprises no apertures, holes, opening orother penetrations that are either open or to which pipes or otherconduits are attached. All penetrations (such as the primary coolantinlet and outlet ports 103, 104) in the reactor pressure vessel 100 arelocated in the above-ground minor portion 131, and more specifically inthe top portion 132 of the body 109 of the reactor pressure vessel 100.In one embodiment, it is further preferred that the major portion 130 bea unitary construct with no connections, joints, or welds.

The bottom portion 115 of the reactor pressure vessel 100 is laterallyrestrained by a lateral seismic restraint system 160 that spans thespace between the body 109 of the reactor pressure vessel 100 and thewall surfaces 411 of the reactor well 410 to withstand seismic events.The seismic restraint system 160, which comprises a plurality ofresiliently compressible struts 161, allows for free axial and diametralthermal expansion of the reactor vessel. The bottom of the reactor well410 contains engineered features to flood it with water to providedefense-in-depth against a (hypothetical, non-mechanistic) accident thatproduces a rapid rise in the enthalpy of the reactor's contents. Becausethe reactor system 1000 is designed to prevent loss of the primarycoolant 101 by leaks or breaks and the reactor well 410 can be floodedat will, burn-through of the reactor pressure vessel 100 by molten fuel(corium) can be ruled out as a credible postulate. This inherently safeaspect simplifies the design and analysis of the reactor system 1000.

Referring now to FIGS. 2 and 4-5 concurrently, an embodiment of the heatexchange sub-system 200 is illustrated. While a specific embodiment ofthe heat exchange sub-system 200 will be described herein, it is to beunderstood that, in alternate embodiments, one or more of components canbe omitted as desired. For example, in certain embodiments, one or bothof the horizontal superheaters 205, 206 may be omitted. In certain otherembodiments, one of the horizontal steam generators 203, 204 may beomitted and/or combined into the other one of the horizontal steamgenerators 203,204. Moreover, additional equipment may be incorporatedas necessary so long as the natural circulation of the primary coolant101 through the closed-loop primary coolant circuit 300 is notprohibited through the introduction of substantial head loss.

As mentioned above, the heat exchange subsystem 200 comprises anincoming hot leg 201 that introduces heated primary coolant into theportion of the closed-loop primary coolant circuit 300 that passesthrough the heat exchange sub-system 200 and an outgoing cold leg 202that removes cooled primary coolant from the portion of the closed-loopprimary coolant circuit 300 that passes through the heat exchangesub-system 200. In order to minimize (and in some embodiments eliminate)pressure loss in the closed-loop primary coolant circuit 300 caused byan increase in the elevation of the primary coolant flow, the steamgenerators 203, 204 and the superheaters 205, 206 are all of thehorizontal genre (i.e., the tubes which carry the primary coolant extendsubstantially horizontal through the shell-side fluid) and are inhorizontal alignment with each other where possible.

Within the heat exchange sub-system 200, the primary coolant flow of theclosed-loop primary coolant circuit 300 is divided into two paths 211,212 at a flow divider 215. The flow divider 210 can be a three-wayvalve, a three-way mass flow controller, or a simple Y plumbing joint.The first path 211, which carries the majority of the primary coolantflow, travels through the first horizontal steam generator 203 and thenthrough the second horizontal steam generator 204. Meanwhile, secondpath 212, which carries a minority of the primary coolant flow, travelsthrough the first horizontal superheater 205 and then through the secondhorizontal superheater 206. After passing through the first and secondhorizontal steam generators 203, 204 and the first and second horizontalsuperheaters 205, 206, the first and second paths 211, 212 converge in aflow converger 216, which combines the primary coolant flows of thefirst and second paths 211, 212 and directs the combined flow to theoutgoing cold leg 202. As with the flow divider 215, the flow converger216 may be a three-way valve, a three-way mass flow controller, or asimple Y plumbing joint.

In one embodiment, 10% to 15% of the incoming primary coolant flow thatenters the heat exchange sub-system 200 via the hot leg 201 is directedinto the second path 212 while the remaining 85% to 90% of the incomingprimary coolant is directed into the first path 211. In one specificexample, the incoming primary coolant that enters the heat exchangesub-system 200 via the hot leg 201 has a flow rate of 5 to 7 millionlbs./hr. In this example, 0.6 to 1 million lbs./hr. of the primarycoolant is directed into the second path 212 while the remainder of theprimary coolant flow is directed into the first path 211.

The first and second horizontal steam generators 203, 204 are operablycoupled in series to one another along the first path 211 of theclosed-loop primary coolant circuit 300. Both of the horizontal steamgenerators 203, 204 are horizontally disposed shell-and-tube heatexchangers. The first horizontal steam generator 203 is a high pressuresteam generator while the second horizontal steam generator 204 is a lowpressure steam generator (in comparison to the high pressure steamgenerator). The high first steam generator 203 is located upstream ofthe second horizontal steam generator 204 along the closed-loop primarycoolant circuit 300. Similarly, the first and second horizontalsuperheaters 295, 296 are operably coupled in series to one anotheralong the second path 212 of the closed loop primary coolant circuit300. The first horizontal superheater 205 is a high pressure superheaterwhile the second horizontal superheater 206 is a low pressuresuperheater (in comparison to the high pressure superheater). The highfirst steam superheater 205 is located upstream of the second horizontalsuperheater 206 along the closed-loop primary coolant circuit 300.Furthermore, the first and second superheaters 205, 206 are located inparallel to the first and second horizontal steam generators 203, 204along the closed-loop primary coolant circuit 300.

Furthermore, the first and second horizontal steam generators 203, 204are interconnected by a return header so that the hot primary coolantentering the first horizontal steam generator 203 heats the secondarycoolant to make steam for the high-pressure turbine 220 and thenproceeds to the second horizontal steam generator 204 with minimalpressure loss to make steam for the low-pressure turbine 221.

The flow of the primary coolant in the first path 211 is used to converta secondary coolant flowing through the shell-side of the first andsecond horizontal steam generators 203, 204 from liquid-phase togas-phase through the transfer of heat form the primary coolant to thesecondary coolant within the first and second horizontal steamgenerators 203, 204. Because the flow of the primary coolant through thefirst and horizontal second steam generators 203, 204 is substantiallyhorizontal in nature, the flow of the primary coolant through the firstpath 211 does not cause any substantial pressure drop in the closed-loopprimary coolant circuit 300 resulting from an increase in elevation.Moreover, because of the horizontal alignment of the first and secondhorizontal steam generators 203, 204 with each other and the primarycoolant outlet and inlet ports 103, 104 of the reactor pressure vessel100 (FIG. 5), the primary coolant flow that travels along the first path211 from the primary coolant outlet port 103 of the reactor pressurevessel 100 to the primary coolant inlet port 104 of the reactor pressurevessel 100 does not cause any substantial pressure drop in theclosed-loop primary coolant circuit 300 resulting from an increase inelevation. While the achievement of substantial zero pressure drop inthe closed-loop primary coolant circuit 300 resulting from an increasein elevation is exemplified in terms of a horizontal flow, it ispossible that such substantial zero pressure drop can be achieved by adecline in elevation as the primary coolant flows downstream in theclosed-loop primary coolant circuit 300.

The flow of the primary coolant in the second path 212 is used tosuperheat the vapor-phase of the secondary coolant exiting the first andsecond horizontal steam generators 203, 204 via the first and secondhorizontal superheaters 205, 206 respectively, thereby further dryingthe vapor-phase of the secondary coolant. The use if the horizontalsuperheaters enhance the thermodynamic efficiency of the turbine cycle,carried out on the high pressure turbine 220 and the low pressureturbine 221.

The first and second horizontal superheaters 205, 206 are horizontallydisposed shell-and-tube heat exchanger positioned directly above (and inseries) with the first and second steam generators 203, 204 (FIG. 5).However, due to the slight increase in the elevation of the superheaters205, 206 resulting from their location above the first and secondhorizontal steam generators 203, 204, the flow of the primary coolant inthe second path 212 does cause some pressure drop in the closed-loopprimary coolant circuit 300 resulting from an increase in elevation.However, because only a small amount (10% to 15%) of the total primarycoolant that flows through the heat exchange subsystem 200 is directedinto the second path 212 and through the horizontal superheaters 205,206, the pressure drop does not significantly affect the desired naturalcirculation. Moreover, the increase in elevation is negligible whencompared to the height of the flow driving water columns. In such anembodiment, at least 85% of the flow of the primary coolant through theheat exchange sub-system 200 is still entirely horizontal from theprimary coolant outlet 103 to the primary coolant inlet 104 and does notcause any substantial pressure drop in the closed-loop primary coolantcircuit 300 due to increase in elevation. Further, in certain alternateembodiments, the horizontal superheaters 205, 206 could be eliminatedand/or repositioned to be in horizontal alignment with the horizontalsteam generators 203, 204.

As shown in FIG. 5, the first and second horizontal steam generators203, 204 are coupled directly to the each other and to the reactorpressure vessel 100. More specifically, the inlet of the firsthorizontal steam generator 203 is coupled directly to the primarycoolant outlet port 103 of the reactor pressure vessel 100 while theoutlet of the first horizontal steam generator 203 is coupled directlyto the inlet of the second horizontal steam generator 204. The outlet ofthe second horizontal steam generator 204, is in turn, coupled directlyto the primary coolant inlet port 104 of the reactor pressure vessel100. The first and second horizontal steam generators 203, 204 arearranged so as to extend substantially parallel to one another, therebycollectively forming a generally U-shaped structure. Thus, the firstpath 211 also takes on a generally U-shape. In certain embodiments, thefirst and second horizontal steam generators 203, 204 are integrallywelded to the reactor vessel 100 and to each other.

Referring now to FIGS. 2 and 3A-B, each of the first and secondhorizontal steam generators 203, 204 comprise a preheating zone 208, 210and a boiling zone 207, 209. Both of the first and second horizontalsteam generators 203, 204 are of the single-pass type in which theprimary coolant flow of the first path 211 is the tube-side fluid. Eachof the single-pass tubes 330 extend substantially horizontally throughthe preheating zones 208, 210 and the boiling zones 207, 209. Thesecondary coolant circuit has a main feedwater intake 501 and a returnto condenser exit 502 into and out of the heat exchange sub system 200respectively.

The secondary coolant, which is in the liquid-phase 505, enters each ofthe first and second horizontal steam generators 203, 204 along line503. The incoming liquid phase 505 of the secondary coolant is preheatedwithin the preheater zones 208, 210 of the first and second horizontalsteam generators 203, 204. The secondary coolant in liquid-phase 505flows through a tortuous path as shell-side fluid in the preheater zones208, 210 and then enters the boiling zones 207, 209, where it is furtherheated by the primary coolant flow passing through the tubes 330. In theboiling zones 207, 209, the liquid-phase secondary coolant 505 vaporizesand exits the first and second horizontal steam generators 203, 204 ashigh pressure and low pressure steam 504 that is respectively suppliedto the high and low pressure turbines 220, 221.

The shells of the horizontal steam generators 203, 204 and thehorizontal superheaters 205, 206 provide additional barriers againstpotential large break LOCAs, as do the turning plenum and the eccentricflanges that join the steam generators 203, 204 to the reactor pressurevessel 100, as shown in FIGS. 4 and 5. All systems connected to thereactor vessel 100 use a similar approach to ensure that there is nopotential for a large-break LOCA that could rapidly drain the water fromthe reactor vessel 100 and uncover the reactor core 102. As long as thereactor core 102 is covered under all potential conditions of operationand hypothetical accident, the release of radioactive material to thepublic is minimal.

As explained in the foregoing, the reactor system 1000 is anintrinsically safe reactor which, in the event of a problem external tothe reactor containment building or within containment, is designed toautomatically shut down in a safe mode with natural circulation cooling.Nevertheless, to instill maximum confidence, a number of redundantsafety systems can be engineered to protect public health and safetyunder hypothetical accident scenarios that are unknown or unknowable,i.e., cannot be mechanistically postulated. In the case of an abnormalcondition when the normal heat transport path through the steamgenerators are not available, then the pressure in the reactor vessel100 will begin to increase. In such a case rupture discs will breachallowing the reactor coolant to flow into a kettle reboiler locatedoverhead. The kettle will have a large inventory of water that willserve to extract the heat from the reactor coolant until the systemshuts down. Diverse systems perform duplicate or overlapping functionsusing different physical principles and equipment to ensure that acommon-mode failure is impossible.

As used throughout, ranges are used as shorthand for describing each andevery value that is within the range. Any value within the range can beselected as the terminus of the range. In addition, all references citedherein are hereby incorporated by referenced in their entireties. In theevent of a conflict in a definition in the present disclosure and thatof a cited reference, the present disclosure controls.

While the invention has been described with respect to specific examplesincluding presently preferred modes of carrying out the invention, thoseskilled in the art will appreciate that there are numerous variationsand permutations of the above described systems and techniques. It is tobe understood that other embodiments may be utilized and structural andfunctional modifications may be made without departing from the scope ofthe present invention. Thus, the spirit and scope of the inventionshould be construed broadly as set forth in the appended claims.

What is claimed is:
 1. A natural circulation nuclear reactor systemcomprising: a reactor pressure vessel having a body comprising aninternal cavity containing a primary coolant, a first portion of thereactor pressure vessel located above a ground level and a secondportion of the reactor pressure vessel located below the ground level; apartition dividing the internal cavity of the reactor pressure vesselinto a primary coolant riser passageway and a primary coolant downcomerpassageway, wherein the partition is configured to prohibit cross-flowof the primary coolant between the primary coolant downcomer passagewayand the primary coolant riser passageway at a top portion of the reactorpressure vessel; a reactor core comprising nuclear fuel disposed withinthe primary coolant riser passageway of the internal cavity at a bottomportion of the reactor pressure vessel; the reactor pressure vesselcomprising a primary coolant outlet port in fluid communication with atop portion of the primary coolant riser passageway and a primarycoolant inlet port in fluid communication with a top portion of theprimary coolant downcomer passageway; a heat exchange sub-system locatedoutside of the reactor pressure vessel, the primary coolant outlet portfluidly coupled to the heat exchange sub-system to form an incoming hotleg of the heat exchange sub-system and the primary coolant inlet portfluidly coupled to the heat exchange sub-system to form an outgoing coldleg of the heat exchange sub-system; a closed-loop primary coolantcircuit that flows the primary coolant through the reactor pressurevessel to cool the reactor core and through the heat exchange sub-systemto transfer heat to a secondary coolant; and wherein operation of thereactor core causes natural circulation of the primary coolant throughthe closed-loop primary coolant circuit in a single phase.
 2. Thenatural circulation nuclear reactor system according to claim 1 whereinthe partition comprises a first portion hating a first end that iscoupled to an inner surface of the reactor pressure vessel and a secondportion that extends from the first portion to a terminal end of thepartition, the terminal end of the partition spaced front a floor of thereactor pressure vessel, a bottom plenum formed between the terminal endof the partition and the floor of the reactor pressure vessel thatallows cross-flow of the primary coolant from the primary coolantdowncomer passageway to the primary coolant riser passageway, whereinoperation of the reactor core causes the primary coolant to rise withinthe primary coolant riser passageway.
 3. The natural circulation nuclearreactor system according to claim 2 wherein the primary coolant withinthe primary coolant riser passageway is fluidly coupled to the primarycoolant within the primary coolant downcomer passageway only at thebottom plenum.
 4. The natural circulation nuclear reactor systemaccording to claim 2 wherein the primary coolant outlet port is locatedat an elevation above the first end of the first portion of thepartition and wherein the primary coolant inlet port is located at anelevation below the first end of the first portion of the partition. 5.The natural circulation nuclear reactor system according to claim 1wherein the primary coolant outlet port and the primary coolant inletport are at substantially the same elevation.
 6. The natural circulationnuclear reactor system according to claim 1 wherein the partition is amulti-layer construction comprising an insulating layer that issandwiched between two outer layers, wherein the insulating layercomprises a refractory material and the two outer layers comprisestainless steel.
 7. The natural circulation nuclear reactor systemaccording to claim 1 wherein the partition has an effective coefficientof thermal conductivity measured from the primary coolant riserpassageway to the primary coolant downcomer passageway that is less thanan effective coefficient of thermal conductivity of the primary coolant.8. The natural circulation nuclear reactor system according to claim 1wherein the partition comprises a first wall portion that is spacedapart from an inner surface of the reactor pressure vessel and a secondwall portion that extends from the first wall portion to the innersurface of the reactor pressure vessel, the first wall portion having aninner surface and an outer surface, the inner surface of the first wallportion forming the primary coolant riser passageway, and the primarycoolant downcomer passageway being formed between the outer surface offirst wall portion and the inner surface of the reactor pressure vesselwith the second wall portion forming a roof of the primary coolantdowncomer passageway, the primary coolant downcomer passageway being anannular passageway that circumferentially surrounds the primary coolantriser passageway.
 9. The natural circulation nuclear reactor systemaccording to claim 1 wherein the reactor pressure vessel extends along asubstantially vertical axis, wherein the second portion of the reactorpressure vessel comprises a major portion of an axial length of thereactor pressure vessel, wherein the reactor core is located below theground level, and wherein the primary coolant outlet port, the primarycoolant inlet port, and the heat exchange sub-system are located abovethe ground level.
 10. The natural circulation nuclear reactor systemaccording to claim 9 wherein the major portion of the reactor pressurevessel is at least 75% of the axial length of the reactor pressurevessel and is free of penetrations.
 11. The natural circulation nuclearreactor system according to claim 1 further comprising a reactor flangewelded to the first portion of the reactor pressure vessel, the reactorflange supporting the weight of the reactor pressure vessel in avertically oriented cantilevered manner with the body of the reactorpressure vessel spaced apart from wall surfaces and a floor surface of areactor well within which the reactor pressure vessel is located,thereby allowing the reactor pressure vessel to radially and axiallyexpand as the reactor core heats up during operation
 12. The naturalcirculation nuclear reactor system according to claim 11 wherein thereactor flange includes a first flange, a second flange located belowthe first flange and resting on the ground level, and a plurality ofwelded lugs extending vertically between the first and second flanges,and wherein the primary coolant inlet and outlet ports are locatedbetween the first and second flanges.
 13. The natural circulationnuclear reactor system according to claim 1 wherein the naturalcirculation nuclear reactor system is free of active equipment forforcing circulation of the primary coolant through the closed-loopprimary coolant circuit.
 14. A natural circulation nuclear reactorsystem comprising: an elongated reactor pressure vessel having aninternal cavity containing a primary coolant, the reactor pressurevessel extending along a substantially vertical axis, a major portion ofthe axial length of the reactor pressure vessel located below a groundlevel; a reactor core comprising nuclear fuel disposed within theinternal cavity at a bottom portion of the reactor pressure vessel andbelow the ground level; the reactor pressure vessel comprising a primarycoolant outlet port located above the ground level; the reactor pressurevessel comprising a primary coolant inlet port located above the groundlevel; a heat exchange sub-system located outside of the reactorpressure vessel and above the ground level, an incoming hot leg of theheat exchange sub-system fluidly coupled to the primary coolant outletport and an outgoing cold leg of the heat exchange sub-system fluidlycoupled to the primary coolant inlet port; and wherein the major portionof the reactor pressure vessel is free of penetrations.
 15. The naturalcirculation nuclear reactor system according to claim 14 furthercomprising: a partition dividing the internal cavity of the reactorpressure vessel into a primary coolant riser passageway and a primarycoolant downcomer passageway, the reactor core disposed within theprimary coolant riser passageway; the primary coolant outlet port influid communication with a top portion of the primary coolant riserpassageway; and the primary coolant inlet port in fluid communicationwith a top portion of the primary coolant downcomer passageway.
 16. Thenatural circulation nuclear reactor system according to claim 15 whereinthe partition comprises a first wall portion that is spaced apart froman inner surface of the reactor pressure vessel and a second wallportion that extends from the first wall portion to the inner surface ofthe reactor pressure vessel, the first wall portion having an innersurface and an outer surface, the inner surface of the first wallportion forming the primary coolant riser passageway; and the, primarycoolant downcomer passageway being formed between the outer surface offirst wall portion and the inner surface of the reactor pressure vesselwith the second wall portion forming a roof of the primary coolantdowncomer passageway, the primary coolant downcomer passageway being anannular passageway that circumferentially surrounds the primary coolantriser passageway.
 17. The natural circulation nuclear reactor systemaccording to claim 14 wherein the heat exchange sub-system and theinternal cavity of the reactor pressure vessel collectively form aclosed-loop primary coolant circuit that flows the primary coolantthrough the reactor pressure vessel to cool the reactor core and throughthe beat exchange sub-system to transfer heat to a secondary coolant,wherein operation of the reactor core causes natural circulation. of theprimary coolant through the closed-loop primary coolant circuit in asingle phase, wherein the secondary coolant is converted fromliquid-phase to gas-phase within the heat exchange sub-system by theheat transferred from the primary coolant, and wherein the flow of theprimary coolant within the heat exchange sub-system that converts thesecondary coolant from liquid-phase to gas-phase does not cause anysubstantial pressure drop in the closed-loop primary coolant circuit.18. A natural circulation nuclear reactor system comprising: anelongated reactor pressure vessel extending along a substantiallyvertical axis and having an internal cavity containing a primarycoolant; a reactor core comprising nuclear fuel disposed within theinternal cavity at a bottom portion of the reactor pressure vessel; apartition dividing the internal cavity of the reactor pressure vesselinto a primary coolant riser passageway and a primary coolant downcomerpassageway, the reactor core disposed within the primary coolant riserpassageway; the reactor pressure vessel comprising a primary coolantoutlet port in fluid communication with a top portion of the primarycoolant riser passageway; the reactor pressure vessel comprising aprimary coolant inlet port in fluid communication with a top portion ofthe primary coolant downcomer riser passageway; and at least one steamgenerator located outside of the reactor pressure vessel, an incominghot leg of the steam generator fluidly coupled to the primary coolantoutlet port and an outgoing cold leg of the steam generator fluidlycoupled to the primary coolant inlet port.
 19. The natural circulationnuclear reactor system according to claim 19 wherein the partitioncomprises a first wall portion, extending in a direction of the verticalaxis, that is spaced apart from an inner surface of the reactor pressurevessel and a second wall portion that extends from the first wallportion to the inner surface of the reactor pressure vessel, the firstwall portion having an inner surface and an outer surface, the innersurface of the first wall portion forming the primary coolant riserpassageway, and the primary coolant downcomer passageway being formedbetween the outer surface of first wall portion and the inner surface ofthe reactor pressure vessel, the primary coolant downcomer passagewaybeing an annular passageway that circumferentially surrounds the primarycoolant riser passageway.
 20. The natural circulation nuclear reactorsystem of claim 19 wherein the second wall of the partition contacts theinner surface of the reactor pressure vessel at a first axial heightabove ground level, wherein the primary coolant outlet port is locatedat a second axial height above ground level, and wherein the primarycoolant inlet port is located at a third axial height above groundlevel, wherein the second axial height is greater than the first axialheight and the first axial height is greater than the third axialheight.